[Federal Register Volume 74, Number 14 (Friday, January 23, 2009)]
[Notices]
[Pages 4247-4257]
From the Federal Register Online via the Government Publishing Office [www.gpo.gov]
[FR Doc No: E9-1152]
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NUCLEAR REGULATORY COMMISSION
[NRC-2009-0004]
Notice; Applications and Amendments to Facility Operating
Licenses Involving Proposed No Significant Hazards Considerations and
Containing Sensitive Unclassified Non-Safeguards Information or
Safeguards Information and Order Imposing Procedures for Access to
Sensitive Unclassified Non-Safeguards Information or Safeguards
Information
I. Background
Pursuant to section 189a.(2) of the Atomic Energy Act of 1954, as
amended (the Act), the U.S. Nuclear Regulatory Commission (the
Commission or NRC) staff is publishing this notice. The Act requires
the Commission publish notice of any amendments issued, or proposed to
be issued and grants the Commission the authority to issue and make
immediately effective any amendment to an operating license upon a
determination by the Commission that such amendment involves no
significant hazards consideration, notwithstanding the pendency before
the Commission of a request for a hearing from any person.
This notice includes notices of amendments containing sensitive
unclassified non-safeguards information (SUNSI) or safeguards
information (SGI).
Notice of Consideration of Issuance of Amendments to Facility Operating
Licenses, Proposed No Significant Hazards Consideration Determination,
and Opportunity for a Hearing
The Commission has made a proposed determination that the following
amendment requests involve
[[Page 4248]]
no significant hazards consideration. Under the Commission's
regulations in 10 CFR 50.92, this means that operation of the facility
in accordance with the proposed amendment would not (1) involve a
significant increase in the probability or consequences of an accident
previously evaluated; or (2) create the possibility of a new or
different kind of accident from any accident previously evaluated; or
(3) involve a significant reduction in a margin of safety. The basis
for this proposed determination for each amendment request is shown
below.
The Commission is seeking public comments on this proposed
determination. Any comments received within 30 days after the date of
publication of this notice will be considered in making any final
determination.
Normally, the Commission will not issue the amendment until the
expiration of 60 days after the date of publication of this notice. The
Commission may issue the license amendment before expiration of the 60-
day period provided that its final determination is that the amendment
involves no significant hazards consideration. In addition, the
Commission may issue the amendment prior to the expiration of the 30-
day comment period should circumstances change during the 30-day
comment period such that failure to act in a timely way would result,
for example in derating or shutdown of the facility. Should the
Commission take action prior to the expiration of either the comment
period or the notice period, it will publish in the Federal Register a
notice of issuance. Should the Commission make a final No Significant
Hazards Consideration Determination, any hearing will take place after
issuance. The Commission expects that the need to take this action will
occur very infrequently.
Written comments may be submitted by mail to the Chief, Rulemaking,
Directives and Editing Branch, TWB-05-B01M, Division of Administrative
Services, Office of Administration, U.S. Nuclear Regulatory Commission,
Washington, DC 20555-0001, and should cite the publication date and
page number of this Federal Register notice. Documents may be examined,
and/or copied for a fee, at the NRC(s Public Document Room (PDR),
located at One White Flint North, Public File Area O1 F21, 11555
Rockville Pike (first floor), Rockville, Maryland.
Within 60 days after the date of publication of this notice, any
person(s) whose interest may be affected by this action may file a
request for a hearing and a petition to intervene with respect to
issuance of the amendment to the subject facility operating license.
Requests for a hearing and a petition for leave to intervene shall be
filed in accordance with the Commission's ``Rules of Practice for
Domestic Licensing Proceedings'' in 10 CFR Part 2. Interested person(s)
should consult a current copy of 10 CFR 2.309, which is available at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland, or
at http://www.nrc.gov/reading-rm/doc-collections/cfr/part002/part002-0309.html. Publicly available records will be accessible from the
Agencywide Documents Access and Management System's (ADAMS) Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm.html. If a request for a hearing or petition for
leave to intervene is filed within 60 days, the Commission or a
presiding officer designated by the Commission or by the Chief
Administrative Judge of the Atomic Safety and Licensing Board Panel,
will rule on the request and/or petition; and the Secretary or the
Chief Administrative Judge of the Atomic Safety and Licensing Board
will issue a notice of a hearing or an appropriate order.
As required by 10 CFR 2.309, a petition for leave to intervene
shall set forth with particularity the interest of the petitioner in
the proceeding, and how that interest may be affected by the results of
the proceeding. The petition should specifically explain the reasons
why intervention should be permitted with particular reference to the
following general requirements: (1) The name, address, and telephone
number of the requestor or petitioner; (2) the nature of the
requestor's/petitioner's right under the Act to be made a party to the
proceeding; (3) the nature and extent of the requestor's/petitioner's
property, financial, or other interest in the proceeding; and (4) the
possible effect of any decision or order which may be entered in the
proceeding on the requestor's/petitioner's interest. The petition must
also set forth the specific contentions which the petitioner/requestor
seeks to have litigated at the proceeding.
Each contention must consist of a specific statement of the issue
of law or fact to be raised or controverted. In addition, the
petitioner/requestor shall provide a brief explanation of the bases for
the contention and a concise statement of the alleged facts or expert
opinion which support the contention and on which the petitioner/
requestor intends to rely in proving the contention at the hearing. The
petitioner/requestor must also provide references to those specific
sources and documents of which the petitioner is aware and on which the
petitioner/requestor intends to rely to establish those facts or expert
opinion. The petition must include sufficient information to show that
a genuine dispute exists with the applicant on a material issue of law
or fact. Contentions shall be limited to matters within the scope of
the amendment under consideration. The contention must be one which, if
proven, would entitle the petitioner/requestor to relief. A petitioner/
requestor who fails to satisfy these requirements with respect to at
least one contention will not be permitted to participate as a party.
Those permitted to intervene become parties to the proceeding,
subject to any limitations in the order granting leave to intervene,
and have the opportunity to participate fully in the conduct of the
hearing.
If a hearing is requested, and the Commission has not made a final
determination on the issue of no significant hazards consideration, the
Commission will make a final determination on the issue of no
significant hazards consideration. The final determination will serve
to decide when the hearing is held. If the final determination is that
the amendment request involves no significant hazards consideration,
the Commission may issue the amendment and make it immediately
effective, notwithstanding the request for a hearing. Any hearing held
would take place after issuance of the amendment. If the final
determination is that the amendment request involves a significant
hazards consideration, any hearing held would take place before the
issuance of any amendment.
All documents filed in NRC adjudicatory proceedings, including a
request for hearing, a petition for leave to intervene, any motion or
other document filed in the proceeding prior to the submission of a
request for hearing or petition to intervene, and documents filed by
interested governmental entities participating under 10 CFR 2.315(c),
must be filed in accordance with the NRC E-Filing rule, which the NRC
promulgated in August 28, 2007 (72 FR 49139). The E-Filing process
requires participants to submit and serve all adjudicatory documents
over the internet, or in some cases to mail copies on electronic
storage media. Participants may not submit paper copies of their
filings unless they seek
[[Page 4249]]
a waiver in accordance with the procedures described below.
To comply with the procedural requirements of E-Filing, at least
ten (10) days prior to the filing deadline, the petitioner/requestor
must contact the Office of the Secretary by e-mail at
[email protected], or by calling (301) 415-1677, to request (1) a
digital ID certificate, which allows the participant (or its counsel or
representative) to digitally sign documents and access the E-Submittal
server for any proceeding in which it is participating; and/or (2)
creation of an electronic docket for the proceeding (even in instances
in which the petitioner/requestor (or its counsel or representative)
already holds an NRC-issued digital ID certificate). Each petitioner/
requestor will need to download the Workplace Forms
Viewer(tm) to access the Electronic Information Exchange
(EIE), a component of the E-Filing system. The Workplace Forms
Viewer(tm) is free and is available at http://www.nrc.gov/site-help/e-submittals/install-viewer.html. Information about applying
for a digital ID certificate is available on NRC's public Web site at
http://www.nrc.gov/site-help/e-submittals/apply-certificates.html.
Once a petitioner/requestor has obtained a digital ID certificate,
had a docket created, and downloaded the EIE viewer, it can then submit
a request for hearing or petition for leave to intervene. Submissions
should be in Portable Document Format (PDF) in accordance with NRC
guidance available on the NRC public Web site at http://www.nrc.gov/site-help/e-submittals.html. A filing is considered complete at the
time the filer submits its documents through EIE. To be timely, an
electronic filing must be submitted to the EIE system no later than
11:59 p.m. Eastern Time on the due date. Upon receipt of a
transmission, the E-Filing system time-stamps the document and sends
the submitter an e-mail notice confirming receipt of the document. The
EIE system also distributes an e-mail notice that provides access to
the document to the NRC Office of the General Counsel and any others
who have advised the Office of the Secretary that they wish to
participate in the proceeding, so that the filer need not serve the
documents on those participants separately. Therefore, applicants and
other participants (or their counsel or representative) must apply for
and receive a digital ID certificate before a hearing request/petition
to intervene is filed so that they can obtain access to the document
via the E-Filing system.
A person filing electronically may seek assistance through the
``Contact Us'' link located on the NRC Web site at http://www.nrc.gov/site-help/e-submittals.html or by calling the NRC electronic filing
Help Desk, which is available between 8 a.m. and 8 p.m., Eastern Time,
Monday through Friday. The electronic filing Help Desk can be contacted
by telephone at 1-866-672-7640 or by e-mail at [email protected].
Participants who believe that they have a good cause for not
submitting documents electronically must file a motion, in accordance
with 10 CFR 2.302(g), with their initial paper filing requesting
authorization to continue to submit documents in paper format. Such
filings must be submitted by: (1) FIRst class mail addressed to the
Office of the Secretary of the Commission, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0001, Attention: Rulemaking and
Adjudications Staff; or (2) courier, express mail, or expedited
delivery service to the Office of the Secretary, Sixteenth Floor, One
White Flint North, 11555 Rockville Pike, Rockville Maryland 20852,
Attention: Rulemaking and Adjudications Staff. Participants filing a
document in this manner are responsible for serving the document on all
other participants. Filing is considered complete by first-class mail
as of the time of deposit in the mail, or by courier, express mail, or
expedited delivery service upon depositing the document with the
provider of the service.
Non-timely requests and/or petitions and contentions will not be
entertained absent a determination by the Commission or the presiding
officer of the Atomic Safety and Licensing Board that the petition and/
or request should be granted and/or the contentions should be admitted,
based on a balancing of the factors specified in 10 CFR 2.309(c)(1)(i)-
(viii).
Documents submitted in adjudicatory proceedings will appear in
NRC's electronic hearing docket which is available to the public at
http://ehd.nrc.gov/ehd_proceeding/home.asp, unless excluded pursuant
to an order of the Commission, an Atomic Safety and Licensing Board, or
a Presiding Officer. Participants are requested not to include personal
privacy information, such as social security numbers, home addresses,
or home phone numbers in their filings. With respect to copyrighted
works, except for limited excerpts that serve the purpose of the
adjudicatory filings and would constitute a Fair Use application,
participants are requested not to include copyrighted materials in
their submission.
For further details with respect to this amendment action, see the
application for amendment which is available for public inspection at
the Commission's PDR, located at One White Flint North, Public File
Area 01F21, 11555 Rockville Pike (first floor), Rockville, Maryland.
Publicly available records will be accessible from the ADAMS Public
Electronic Reading Room on the Internet at the NRC Web site, http://www.nrc.gov/reading-rm/adams.html. If you do not have access to ADAMS
or if there are problems in accessing the documents located in ADAMS,
contact the PDR Reference staff at 1 (800) 397-4209, (301) 415-4737 or
by e-mail to [email protected].
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Date of amendment request: November 20, 2008.
Description of amendment request: This amendment request contains
sensitive unclassified non-safeguards information (SUNSI). The proposed
amendment revises Technical Specification (TS) 5.6.5, ``Core Operating
Limits Report (COLR),'' to add a reference to an analytical method that
will be used to determine core operating limits. The new reference,
NEDC-33383P, ``GEXL97 Correlation Applicable to ATRIUM-10 Fuel,'' will
allow the licensee to use a Global Nuclear Fuel method to determine
fuel assembly critical power of AREVA ATRIUM-10 fuel.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
Core operating limits are established each operating cycle in
accordance with TS 3.2, ``Power Distribution'' and TS 5.6.5, ``Core
Operating Limits Report (COLR).'' These core operating limits ensure
that the fuel design limits are not exceeded during any conditions
of normal operation or in the event of any Anticipated Operational
Occurrence (AOO). The methods used to determine the operating limits
are those previously found acceptable by the NRC and listed in TS
section 5.6.5.b.
A change to TS 5.6.5.b is requested to include an additional
reference to the list of analytical methods. RBS [River Bend
Station] currently operates with a full core of AREVA ATRIUM-10 fuel
but is scheduled to load GE14 fuel during the next refueling outage.
[[Page 4250]]
RBS plans to use the analysis methods of the new fuel vendor, GNF
[Global Nuclear Fuel], for the analysis of the mixed core. The
GEXL97 correlation accurately models predicted core behavior and
appropriately determines the overall critical power uncertainty of
this method. In addition, the GEXL97 application range covers the
range of expected operation of the ATRIUM-10 fuel during normal
steady state and transient conditions in the RBS reload cores.
The requested TS changes concern the use of analytical methods
and do not involve any plant modifications or operational changes
that could affect any postulated accident precursors or accident
mitigation systems and do not introduce any new accident initiation
mechanisms. The proposed changes have no effect on the type or
amount of radiation released and [have] no effect on predicted
offsite doses in the event of an accident. Thus, the proposed change
does not affect the probability of an accident previously evaluated
nor does it increase the radiological consequences of any accident
previously evaluated.
Therefore, the proposed change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
The proposed TS changes will not change the design function,
reliability, performance, or operation of any plant systems,
components, or structures. It does not create the possibility of a
new failure mechanism, malfunction, or accident initiators not
considered in the design and licensing bases. Plant operation will
continue to be within the core operating limits that are established
using NRC approved methods that are applicable to the RBS design and
the RBS fuel.
Therefore, the proposed change does not create the possibility
of a new or different kind of accident from any previously
evaluated.
3. Does the proposed change involve a significant reduction in a
margin of safety?
Response: No.
The proposed change adds GEXL97 to the list of analytical
methods in TS 5.6.5.b that can be used to determine core operating
limits. Use of the GEXL97 correlation analytical method provides an
equivalent level of protection as that currently provided. The
change does not alter any method of analysis as described in the NRC
approved versions of GESTAR-II [NEDE-24011-P-A, ``General Electric
Standard Application for Reactor Fuel (GESTAR-II)'']. The proposed
change does not modify the safety limits or setpoints at which
protective actions are initiated, and do not change the requirements
governing operation or availability of safety equipment assumed to
operate to preserve the margin of safety.
Therefore, the proposed change does not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Terence A. Burke, Associate General
Counsel--Nuclear Entergy Services, Inc., 1340 Echelon Parkway, Jackson,
Mississippi 39213.
NRC Branch Chief: Michael T. Markley.
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Date of amendment request: December 16, 2008.
Description of amendment request: This amendment request contains
sensitive unclassified non-safeguards information (SUNSI). This
amendment request would revise the Technical Specifications (TSs)
Section 2.1.2, Safety Limit Minimum Critical Power Ratio (SLMCPR) for
two-loop and single-loop operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. The proposed changes to Technical Specification do not
involve a significant increase in the probability of an accident
previously evaluated.
The proposed Safety Limit MCPR (SLMCPR), and its use to
determine the Operating Cycle 18 thermal limits, have been derived
using NRC approved methods specified in the Reference section of the
Technical Specification Bases Section for 2.0 SAFETY LIMITS. These
methods do not change the method of operating the plant and have no
effect on the probability of an accident initiating event or
transient.
The basis of the SLMCPR is to ensure no mechanistic fuel damage
is calculated to occur if the limit is not violated. The new SLMCPR
preserves the margin to transition boiling, and the probability of
fuel damage is not increased.
Therefore, the proposed changes to Technical Specifications do
not involve an increase in the probability or consequences of an
accident previously evaluated.
2. The proposed changes to Technical Specifications do not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
The proposed changes result only from revised methods of
analysis for the Cycle 18 core reload. These methods have been
reviewed and approved by the NRC, do not involve any new or
unapproved method for operating the facility, and do not involve any
facility modifications. No new initiating events or transients
result from these changes.
Therefore, the proposed changes to technical specifications do
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
3. The proposed changes to Technical Specifications do not
involve a significant reduction in a margin of safety.
The margin of safety as defined in the TS bases will remain the
same. The new SLMCPR was derived using NRC approved methods which
are in accordance with the current fuel design and licensing
criteria. The SLMCPR remains high enough to ensure that greater than
99.9% of all fuel rods in the core will avoid transition boiling if
the limit is not violated, thereby preserving the fuel cladding
integrity.
Therefore, the proposed changes to technical specifications do
not involve a significant reduction in the margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. William C. Dennis, Assistant General
Counsel, Entergy Nuclear Operations, Inc., 400 Hamilton Avenue, White
Plains, NY 10601.
NRC Branch Chief: Mark G. Kowal.
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Date of amendment request: July 25, 2008.
Description of amendment request: This amendment request contains
sensitive unclassified non-safeguards information (SUNSI). The proposed
amendments would revise Technical Specification 3.3.1.1, ``Reactor
Protection System (RPS) Instrumentation,'' Surveillance Requirement
(SR) 3.3.1.1.8 and TS 3.3.1.3, ``Oscillation Power Range Monitor (OPRM)
Instrumentation,'' SR 3.3.1.3.2 to increase the frequency interval
between local power range monitor calibrations from 1000 effective full
power hours (EFPH) to 2000 EFPH for the LaSalle County Station, Units 1
and 2 (LSCS).
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed amendment involve a significant increase in
the probability or consequences of an accident previously evaluated?
[[Page 4251]]
Response: No.
The proposed change is a result of increasing the surveillance
interval of the LPRM [Local Power Range Monitor] calibration
frequency from 1000 EFPH to 2000 EFPH. Increasing the frequency
interval between required LRPM calibrations is acceptable due to
improvements in the fuel analytical bases and therefore, the revised
surveillance interval continues to ensure that the LPRM detector
signal is adequately calibrated. Extending the LPRM calibration
surveillance interval will increase the LPRM signal uncertainty
value used in the LSCS SLMCPR [Safety Limit Minimum Critical Power
Ratio] analysis, however, this increase in the LRPM signal
uncertainty value is acceptable since the increase is bounded by the
values used by the AREVA analysis.
This change will not alter the operation of process variables,
structures, systems, or components as described in the LSCS Updated
Final Safety Analysis Report (UFSAR). The proposed change does not
alter the initiation conditions or operational parameters for the
system and there is no new equipment introduced by the extension of
the LPRM calibration frequency interval. The performance of the
Average Power Range Monitor (APRM), Rod Block Monitor (RBM) and
Oscillation Power Range Monitor (OPRM) systems are not significantly
affected by the proposed surveillance interval increase. The
proposed LPRM calibration interval extension will have no
significant effect on the Reactor Protection System (RPS)
instrumentation accuracy during power maneuvers or transients and
will, therefore, not significantly affect the performance of the
RPS. As such, the probability of occurrences for a previously
evaluated accident is not increased.
The radiological consequences of an accident can be affected by
the thermal limits existing at the time of the postulated accident,
however, increasing the surveillance interval frequency will not
increase the calculated thermal limits since all uncertainties
associated with the increased interval are currently implemented and
are currently used to calculate the existing Safety Limits. Plant
specific evaluation of LPRM sensitivity to exposure has determined
that the extended calibration frequency increases the LPRM signal
uncertainty value used in the LSCS SLMCPR analysis, however, the
increase is bounded by the values currently used in the safety
analysis. Therefore, the thermal limit calculation is not
significantly affected by LPRM calibration frequency, and thus the
radiological consequences of any accident previously evaluated are
not increased.
Based on the above information, the proposed change does not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The performance of the APRM, RBM, and OPRM systems are not
significantly affected by the proposed LPRM surveillance interval
increase. The proposed change does not affect the control parameters
governing unit operation or the response of plant equipment to
transient conditions. For the proposed LPRM extended calibration
interval frequency all uncertainties remain less than the
uncertainties assumed in the existing thermal limit calculations.
The proposed change does not change or introduce any new equipment,
modes of system operation or failure mechanisms.
Based on the above information, the proposed change does not
create the possibility of a new or different kind of accident from
any previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
The proposed change has no impact on equipment design or
fundamental operation, and there are no changes being made to safety
limits or safety system allowable values that would adversely affect
plant safety as a result of the proposed LPRM surveillance interval
increase. The performance of the APRM, RBM, and OPRM systems are not
significantly affected by the proposed change. The margin of safety
can be affected by the thermal limits existing at the time of the
postulated accident; however, uncertainties associated with LPRM
chamber exposure have no significant effect on the calculated
thermal limits. Plant specific evaluation of LPRM sensitivity to
exposure has determined that the extended calibration frequency
increases the LPRM signal uncertainty value used in the LSCS SLMCPR
analysis, however, the increase is bounded by the values currently
used in the safety analysis. The thermal limit calculation is not
significantly affected since the LPRM sensitivity with exposure is
well defined. LPRM accuracy remains within the total nodal power
uncertainty assumed in the thermal analysis basis, therefore
maintaining thermal limits and the safety margin. The proposed
change does not affect safety analysis assumptions or initial
conditions and the margin of safety in the original safety analysis
are therefore maintained.
Based on this information, the proposed change does not involve
a significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
requested amendments involve no significant hazards consideration.
Attorney for licensee: Mr. Bradley J. Fewell, Associate General
Counsel, Exelon Nuclear, 4300 Winfield Road, Warrenville, IL 60555.
NRC Branch Chief: Russell Gibbs.
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Date of amendment request: October 13, 2008.
Description of amendment request: This amendment request contains
sensitive unclassified non-safeguards information (SUNSI). The proposed
amendment would revise the licensing basis by approving adoption of the
Alternative Source Term (AST), in accordance with 10 CFR 50.67, for use
in calculating the loss-of-coolant accident (LOCA) dose consequences.
The proposed amendment would revise the Technical Specifications (TSs)
to (1) change the TS definition for DOSE EQUIVALENT I-131 to adopt
Federal Guidance Report (FGR) 11 dose conversion factors, (2) require
operability of the Standby Liquid Control (SLC) system in Mode 3, to
reflect its credit in the LOCA analysis, (3) establish a Main Steam
(MS) Pathway leakage limit that effectively increases the previous MS
isolation valve leakage limit, and (4) change TS Section 5.5.12 to
reflect a requested permanent exemption from the requirements of 10 CFR
Part 50, Appendix J, Option B, Paragraph III.A, to allow exclusion of
MS Pathway leakage from the overall integrated leakage rate measured
during the performance of a Type A test, and from the requirements of
Appendix J, Option B, Paragraph III.B, to allow exclusion of the MS
Pathway leakage from the combined leakage rate of the penetrations and
valves subject to Type B and C tests.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Do the proposed changes involve a significant increase in the
probability or consequences of an accident previously evaluated?
Response: No.
For the postulated design basis accident (DBA) LOCA, the AST is
an input to the calculations that evaluate the radiological
consequences of a LOCA. The AST and the requested Appendix J
exemption do not affect the design of the plant or the manner in
which the plant is normally operated. Adoption of the AST and the
requested Appendix J exemption do not affect the initiators of a
DBA. Neither the AST nor the requested Appendix J exemption [sic]
affect the response to the DBA LOCA, or the pathway of the radiation
released from the nuclear fuel. Rather, the AST better represents
the physical characteristics of the radiation release.
Because the initiators of a DBA are not affected by adoption of
the AST for LOCA dose assessment, the probability of an accident are
not increased by the proposed amendment or requested Appendix J
exemption.
The AST is an input to calculations used to evaluate the
radiological consequences of
[[Page 4252]]
the LOCA. Use of the AST does not affect the plant response to the
accident, or the pathways to the environment for the radiation and
activity released from the fuel. The LOCA radiological analyses have
been performed using the AST. Adoption of the AST methodology
revises the acceptance criteria for the accident to the limits
specified in 10 CFR 50.67. The results of those analyses demonstrate
that the dose consequences are within the acceptance criteria
presented in 10 CFR 50.67 and in NRC RG [Regulatory Guide] 1.183.
Implementation of the AST for the LOCA involves the use of the
SLC System to control the pH of the suppression pool during
mitigation of a LOCA. As a result the proposed amendment revises the
CNS [Cooper Nuclear Station] TS for the SLC System. These changes do
not require any physical modification of the plant, nor result in
any change in normal plant operation. This additional use of the SLC
system does not compromise or adversely affect the function of the
SLC system as a means of shutting down the reactor in addition to
the control rods.
Therefore, it is concluded that adoption of AST and granting of
the Appendix J exemption do not involve a significant increase in
the consequences of an accident previously evaluated. Based on the
above discussion, it is concluded that the proposed changes do not
involve a significant increase in the probability or consequences of
an accident previously evaluated.
2. Do the proposed changes create the possibility of a new or
different kind of accident from any accident previously evaluated?
Response: No.
Implementation of the LOCA AST and the requested Appendix J
exemption do not involve a physical alteration of the plant or a
change in how the plant is normally operated. No new or different
types of equipment will be installed and there are no physical
modifications to existing equipment associated with the proposed
changes. The proposed changes, effectively increasing the allowable
MSIV leakage, establishing a leakage limit for the MS Pathway, and
crediting the SLC system for LOCA mitigation do not create
initiators or precursors of a new or different kind of accident. New
equipment or personnel failure modes that might initiate a new type
of accident are not created as a result of the proposed amendment.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any previously analyzed.
3. Do the proposed changes involve a significant reduction in a
margin of safety?
Response: No.
The proposed amendment involves the implementation of a new
licensing basis for the design basis LOCA. Approval of this change
from the original source term to an AST, derived in accordance with
the guidance of RG 1.183, results in revised acceptance criteria for
the LOCA analysis. For the LOCA, RG 1.183 sets the Exclusion Area
Boundary (EAB), Low Population Zone (LPZ), and Control Room limit
consistent with 10 CFR 50.67. The AST LOCA radiological analysis has
been performed using conservative methodologies, as specified in RG
1.183. Safety margins have been evaluated and confirmed to have not
been reduced. Analytical conservatism has been utilized to ensure
that the analysis adequately bounds the limiting postulated event.
The dose consequences of the DBA LOCA remain within the acceptance
criteria presented in 10 CFR 50.67 and RG 1.183.
The proposed changes continue to ensure that the doses at the
EAB and LPZ boundary, as well as the Control Room, are within the
corresponding regulatory limits.
Since the proposed amendment continues to ensure the doses at
the EAB, LPZ and Control Room are within corresponding regulatory
limits, the proposed license amendment does not involve a
significant reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on
this review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Mr. John C. McClure, Nebraska Public Power
District, Post Office Box 499, Columbus, NE 68602-0499.
NRC Branch Chief: Michael T. Markley.
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
Date of amendment request: November 5, 2008.
Description of amendment request: This amendment request contains
sensitive unclassified non-safeguards information (SUNSI). The licensee
proposed to increase the current maximum power level authorized by
section 2.C(1) of the renewed facility operating license from 1,775
megawatts thermal (Mwt) to 2,004 Mwt, an approximately 13 percent
increase from the current licensed thermal power. The current maximum
power level of 1,775 Mwt was approved in 1998, an increase of 6.3
percent from the original licensed thermal power of 1670 Mwt. Thus,
when approved, the licensee's proposed amendment would take the maximum
power level to about 20 percent above the original license thermal
power. The licensee's application addresses in details each of the
following major technical areas: extended power uprate, containment
analysis methods change, credit for containment overpressure for low
head emergency core cooling system (ECCS) pumps, and reactor internal
pressure differentials (RIPDs) for the steam dryer.
Basis for proposed no significant hazards consideration
determination: As required by Title 10 of the Code of Federal
Regulations (10 CFR) Part 50.91(a), the licensee has provided its
analysis of the issue of no significant hazards consideration (NSHC).
The licensee's NSHC analysis, addressing each technical area listed
above, is reproduced below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Extended Power Uprate
Response: No.
The probability (frequency of occurrence) of [d]esign [b]asis
[a]ccidents occurring is not affected by the increased power level,
because Monticello Nuclear Generating Plant (MNGP) continues to
comply with the regulatory and design basis criteria established for
plant equipment. A probabilistic risk assessment demonstrates that
the calculated core damage frequencies do not significantly change
due to [e]xtended [p]ower [u]prate (EPU). Scram setpoints (equipment
settings that initiate automatic plant shutdowns) are established
such that there is no significant increase in scram frequency due to
EPU. No new challenges to safety-related equipment result from EPU.
The changes in consequences of postulated accidents, which would
occur from 102 percent of the EPU rated thermal power (RTP) compared
to those previously evaluated, are acceptable. The results of EPU
accident evaluations do not exceed the NRC[-]approved acceptance
limits. The spectrum of postulated accidents and transients has been
investigated, and are shown to meet the plant's currently licensed
regulatory criteria. In the area of fuel and core design, for
example, the Safety Limit Minimum Critical Power Ratio (SLMCPR) and
other applicable Specified Acceptable Fuel Design Limits (SAFDL) are
still met. Continued compliance with the SLMCPR and other SAFDLs
will be confirmed on a cycle[-]specific basis consistent with the
criteria accepted by the NRC.
Challenges to the [r]eactor [c]oolant [p]ressure [b]oundary were
evaluated at EPU conditions (pressure, temperature, flow, and
radiation) and were found to meet their acceptance criteria for
allowable stresses and overpressure margin.
Challenges to the containment have been evaluated, and the
containment and its associated cooling systems continue to meet the
current licensing basis. The increase in the calculated post[-] LOCA
suppression pool temperature above the currently assumed peak
temperature was evaluated and determined to be acceptable.
Radiological release events (accidents) have been evaluated, and
have been shown to meet the guidelines of 10 CFR 50.67.
Containment Analysis Methods Change
Response: No.
The use of passive heat sinks, variable RHR [residual heat
removal] heat exchanger capability K-value, and mechanistic heat and
[[Page 4253]]
mass transfer from the suppression pool surface to the wetwell
airspace after 30 seconds for the long[-]term design[-] basis[-
]accident loss[-]of[-]coolant accident (DBA-LOCA) containment
analysis are not relevant to accident initiation, but rather,
pertain to the method used to accurately evaluate postulated
accidents. The use of these elements does not, in any way, alter
existing fission product boundaries, and provides a conservative
prediction of the containment response to DBA-LOCAs. Therefore, the
containment analysis method change does not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
Credit for Containment Overpressure for Low Head Emergency Core
Cooling System (ECCS) Pumps
Response: No.
These changes update parameters used in the MNGP safety analyses
and expand the range and scope of the analyses. This will result in
a more realistic analysis of available containment overpressure
under design[-]basis accident conditions. The updated analyses
affect only the evaluation of previously reviewed accidents. No
plant structure, system, or component (SSC) is physically affected
by the updated and expanded analyses. No method of operation of any
plant SSC is affected. Therefore, there is no significant increase
in the probability or consequence of a previously evaluated
accident.
Reactor Internal Pressure Differentials (RIPDs) for the Steam Dryer
Response: No.
The revised steam dryer RIPDs are used in evaluating loads in
reactor vessel internals for various conditions (i.e., during
normal, upset and faulted conditions). The values more accurately
represent the actual plant configuration. No plant structure,
system, or component (SSC) is physically affected by the updated and
expanded analyses. No method of operation of any plant SSC is
affected. Therefore, there is no significant increase in the
probability or consequence of a previously evaluated accident.
The analyses supporting the above evaluations were performed at
the EPU power level of 2,004 Mwt.
Therefore, the proposed changes do not involve a significant
increase in the probability or consequences of an accident
previously evaluated.
2. Does the proposed change create the possibility of a new or
different kind of accident from any accident previously evaluated?
Extended Power Uprate
Response: No.
Equipment that could be affected by EPU has been evaluated. No
new operating mode, safety-related equipment lineup, accident
scenario, or equipment failure mode was identified. The full
spectrum of accident considerations has been evaluated and no new or
different kind of accident has been identified. EPU uses developed
technology and applies it within capabilities of existing or
modified plant safety[-]related equipment in accordance with the
regulatory criteria (including NRC[-]approved codes, standards and
methods). No new accidents or event precursors have been identified.
The MNGP TS require revision to implement EPU. The revisions
have been assessed and it was determined that the proposed change
will not introduce a different accident than that previously
evaluated. Therefore, the proposed changes do not create the
possibility of a new or different kind of accident from any accident
previously evaluated.
Containment Analysis Methods Change
Response: No.
The use of passive heat sinks, variable RHR heat exchanger
capability K-value, and mechanistic heat and transfer from the
suppression pool surface to the wetwell airspace after 30 seconds
for the long[-]term DBA-LOCA containment analysis are not relevant
to accident initiation, but pertain to the method used to evaluate
currently postulated accidents. The use of these analytical tools
does not involve any physical changes to plant structures or
systems, and does not create a new initiating event for the spectrum
of events currently postulated. Further, they do not result in the
need to postulate any new accident scenarios. Therefore, the
containment analysis method change does not create the possibility
of a new or different kind of accident from any accident previously
evaluated.
Credit for Containment Overpressure for Low Head ECCS Pumps
Response: No.
The proposed change involves the updating and expansion in scope
of the existing design bases analysis with respect to the available
containment overpressure to cover additional events. No new failure
mode or mechanisms have been created for any plant SSC important to
safety nor has any new limiting single failure been identified as a
result of the proposed analytical changes. Therefore, the change to
containment overpressure credited for low pressure ECCS pumps does
not create the possibility of a new or different kind of accident
from any accident previously evaluated.
Reactor Internal Pressure Differentials for the Steam Dryer
Response: No.
The revised steam dryer RIPDs are used in evaluating loads in
reactor vessel internals for various conditions (i.e., during
normal, upset and faulted conditions). The steam dryer RIPDs are not
relevant to accident initiation, but only pertain to the method used
to evaluate reactor vessel internals loads. The revised steam dryer
RIPD values more accurately represent the actual plant
configuration. Therefore, the change to steam dryer RIPDs does not
create the possibility of a new or different kind of accident from
any accident previously evaluated.
The analyses supporting the above evaluations were performed at
the EPU power level of 2,004 Mwt.
Therefore, the proposed changes do not create the possibility of
a new or different kind of accident from any accident previously
evaluated.
3. Does the proposed change involve a significant reduction in
a margin of safety?
Extended Power Uprate
Response: No.
The EPU affects only design and operational margins. Challenges
to the fuel, reactor coolant pressure boundary, and containment were
evaluated for EPU conditions. Fuel integrity is maintained by
meeting existing design and regulatory limits. The calculated loads
on affected structures, systems and components, including the
reactor coolant pressure boundary, will remain within their design
allowables for design[-]basis event categories. No NRC acceptance
criterion is exceeded. Because the MNGP configuration and responses
to transients and postulated accidents do not result in exceeding
the presently approved NRC acceptance limits, the proposed changes
do not involve a significant reduction in a margin of safety.
Containment Analysis Methods Change
Response: No.
The use of passive heat sinks, variable RHR heat exchanger
capability K-value, and mechanistic heat and mass transfer from the
suppression pool surface to the wetwell airspace after 30 seconds
for the long[-]term DBA-LOCA containment analysis are realistic
phenomena and provide a conservative prediction of the plant
response to DBA-LOCAs. The increase in pressure and temperature are
relatively small and are within design limits. Therefore, the
containment analysis methods change does not involve a significant
reduction in the margin of safety.
Increase in Credit for Containment Overpressure for Low Head ECCS
Pumps
Response: No.
The proposed changes revise containment response analytical
methods and scope for containment pressure to assist in ECCS pump
net positive suction head (NPSH). The changes are still based on
conservative but more realistic analysis of available containment
overpressure determined using analysis methods that minimize
containment pressure and maximize suppression pool temperature.
These changes do not constitute a significant reduction in the
margin of safety.
Reactor Internal Pressure Differentials for the Steam Dryer
Response: No.
The revised steam dryer RIPDs are used in evaluating loads in
reactor vessel internals for various conditions (i.e., during
normal, upset and faulted conditions). The revised steam dryer RIPD
values more accurately represent the actual plant configuration. The
changes are still conservative but more accurately represent the
MNGP configuration. These changes do not constitute a significant
reduction in the margin of safety.
The analyses supporting the above evaluations were performed at
the EPU power level of 2,004 Mwt.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
[[Page 4254]]
The NRC staff has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
proposed amendment involves no significant hazards consideration.
Attorney for licensee: Peter M. Glass, Assistant General Counsel,
Xcel Energy Services, Inc., 414 Nicollet Mall, Minneapolis, MN 55401.
NRC Branch Chief: Lois M. James.
PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric
Station, Unit 2, Luzerne County, Pennsylvania
Date of amendment request: October 30, 2008.
Description of amendment request: This amendment request contains
sensitive unclassified non-safeguards information (SUNSI). This
amendment request would revise PPL Susquehanna, LLC, Unit 2 (PPL)
Technical Specifications (TSs) Section 2.1.1.2, Minimum Critical Power
Ratio Safety Limits (MCPRSLs) for two-loop and single-loop operation.
Basis for proposed no significant hazards consideration
determination: As required by 10 CFR 50.91(a), the licensee has
provided its analysis of the issue of no significant hazards
consideration, which is presented below:
1. Does the proposed change involve a significant increase in
the probability or consequences of an accident previously evaluated?
Response: No.
The proposed change to the two-loop and single-loop MCPRSLs do
not directly or indirectly affect any plant system, equipment,
component, or change the processes used to operate the plant.
Further, the proposed MCPRSLs were generated using NRC approved
methodology and meet the applicable acceptance criteria. Thus, this
proposed amendment does not involve a significant increase in the
probability of occurrence or consequences of an accident previously
evaluated.
Prior to the startup of U2C15, licensing analyses are performed
(using NRC approved methodology referenced in TS Section 5.6.5.b) to
determine changes in the CPR as a result of anticipated operational
occurrences. These results are added to the MCPRSL values to
generate the MCPROLs in the COLR [Core Operating Limits Report].
These limits could be different from those specified for the
previous Unit 2 COLR. The COLR operating limits thus assure that the
MCPRSL will not be exceeded during normal operation or AOOs
[anticipated operational occurrences]. Postulated accidents are also
analyzed prior to the startup and the results shown to be within the
NRC approved criteria.
Therefore, this proposed amendment does not involve a
significant increase in the probability of occurrence or
consequences of an accident previously evaluated.
2. Does the proposed amendment create the possibility of a new
or different kind of accident from any accident previously
evaluated?
Response: No.
The changes to the two-loop and single-loop MCPRSLs do not
directly or indirectly affect any plant system, equipment, or
component and therefore does not affect the failure modes of any of
these items. Thus, the proposed change does not create the
possibility of a previously unevaluated operator error or a new
single failure.
Therefore, this proposed amendment does not create the
possibility of a new or different kind of accident from any
previously evaluated.
3. Does the proposed amendment involve a significant reduction
in a margin of safety?
Response: No.
Since the proposed changes do not alter any plant system,
equipment, component, or processes used to operate the plant, the
proposed change will not jeopardize or degrade the function or
operation of any plant system or component governed by TS. The
proposed two-loop and single-loop MCPRSLs do not involve a
significant reduction in the margin of safety as currently defined
in the Bases of the applicable TS sections, because the proposed
MCPRSLs preserve the required margin of safety.
Therefore, the proposed changes do not involve a significant
reduction in a margin of safety.
The NRC staff has reviewed the licensee's analysis and, based on this
review, it appears that the three standards of 10 CFR 50.92(c) are
satisfied. Therefore, the NRC staff proposes to determine that the
amendment request involves no significant hazards consideration.
Attorney for licensee: Bryan A. Snapp, Esquire, Assoc. General
Counsel, PPL Services Corporation, 2 North Ninth St., GENTW3,
Allentown, PA 18101-1179.
NRC Branch Chief: Mark G. Kowal.
Order Imposing Procedures for Access to Sensitive Unclassified Non-
Safeguards Information (SUNSI) and Safeguards Information (SGI) for
Contention Preparation
Entergy Gulf States Louisiana, LLC, and Entergy Operations, Inc.,
Docket No. 50-458, River Bend Station, Unit 1, West Feliciana Parish,
Louisiana
Entergy Nuclear Operations, Inc., Docket No. 50-293, Pilgrim Nuclear
Power Station, Plymouth County, Massachusetts
Exelon Generation Company, LLC, Docket Nos. 50-373 and 50-374, LaSalle
County Station, Units 1 and 2, LaSalle County, Illinois
Nebraska Public Power District, Docket No. 50-298, Cooper Nuclear
Station, Nemaha County, Nebraska
Nuclear Management Company, LLC, Docket No. 50-263, Monticello Nuclear
Generating Plant, Wright County, Minnesota
PPL Susquehanna, LLC, Docket No. 50-388, Susquehanna Steam Electric
Station, Unit 2, Luzerne County, Pennsylvania
1. This order contains instructions regarding how potential
parties to the proceedings listed above may request access to documents
containing sensitive unclassified information (SUNSI and SGI).
2. Within ten (10) days after publication of this notice of
opportunity for hearing, any potential party as defined in 10 CFR 2.4
who believes access to SUNSI or SGI is necessary for a response to the
notice may request access to SUNSI or SGI. A ``potential party'' is any
person who intends or may intend to participate as a party by
demonstrating standing and the filing of an admissible contention under
10 CFR 2.309. Requests submitted later than ten (10) days will not be
considered absent a showing of good cause for the late filing,
addressing why the request could not have been filed earlier.
3. The requester shall submit a letter requesting permission to
access SUNSI and/or SGI to the Office of the Secretary, U.S. Nuclear
Regulatory Commission, Washington, DC 20555-0001, Attention:
Rulemakings and Adjudications Staff, and provide a copy to the
Associate General Counsel for Hearings, Enforcement and Administration,
Office of the General Counsel, Washington, DC 20555-0001. The expedited
delivery or courier mail address for both offices is U.S. Nuclear
Regulatory Commission, 11555 Rockville Pike, Rockville, MD 20852. The
e-mail address for the Office of the Secretary and the Office of the
[[Page 4255]]
General Counsel are [email protected] and
[email protected], respectively.\1\ The request must
include the following information:
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\1\ See footnote 6. While a request for hearing or petition to
intervene in this proceeding must comply with the filing
requirements of the NRC's ``E-Filing Rule,'' the initial request to
access SUNSI and/or SGI under these procedures should be submitted
as described in this paragraph.
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a. A description of the licensing action with a citation to this
Federal Register notice of opportunity for hearing;
b. The name and address of the potential party and a description of
the potential party's particularized interest that could be harmed by
the action identified in (a);
c. If the request is for SUNSI, the identity of the individual
requesting access to SUNSI and the requester's need for the information
in order to meaningfully participate in this adjudicatory proceeding,
particularly why publicly available versions of the application would
not be sufficient to provide the basis and specificity for a proffered
contention;
d. If the request is for SGI, the identity of the individual
requesting access to SGI and the identity of any expert, consultant or
assistant who will aid the requester in evaluating the SGI, and
information that shows:
(i) Why the information is indispensable to meaningful
participation in this licensing proceeding; and
(ii) The technical competence (demonstrable knowledge, skill,
experience, training or education) of the requester to understand and
use (or evaluate) the requested information to provide the basis and
specificity for a proffered contention. The technical competence of a
potential party or its counsel may be shown by reliance on a qualified
expert, consultant or assistant who demonstrates technical competence
as well as trustworthiness and reliability, and who agrees to sign a
non-disclosure affidavit and be bound by the terms of a protective
order; and
e. If the request is for SGI, Form SF-85, ``Questionnaire for Non-
Sensitive Positions,'' Form FD-258 (fingerprint card), and a credit
check release form completed by the individual who seeks access to SGI
and each individual who will aid the requester in evaluating the SGI.
For security reasons, Form SF-85 can only be submitted electronically,
through a restricted-access database. To obtain online access to the
form, the requester should contact the NRC's Office of Administration
at 301-492-3524.\2\ The other completed forms must be signed in
original ink, accompanied by a check or money order payable in the
amount of $191.00 to the U.S. Nuclear Regulatory Commission for each
individual, and mailed to the: Office of Administration, Security
Processing Unit, Mail Stop TWB-05-B32M, U.S. Nuclear Regulatory
Commission, Washington, DC 20555-0012.
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\2\ The requester will be asked to provide his or her full name,
Social Security number, date and place of birth, telephone number,
and e-mail address. After providing this information, the requester
usually should be able to obtain access to the online form within
one business day.
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These forms will be used to initiate the background check, which
includes fingerprinting as part of a criminal history records check.
Note: copies of these forms do not need to be included with the request
letter to the Office of the Secretary, but the request letter should
state that the forms and fees have been submitted as described above.
4. To avoid delays in processing requests for access to SGI, all
forms should be reviewed for completeness and accuracy (including
legibility) before submitting them to the NRC. Incomplete packages will
be returned to the sender and will not be processed.
5. Based on an evaluation of the information submitted under items
2 and 3.a through 3.d, above, the NRC staff will determine within ten
days of receipt of the written access request whether (1) there is a
reasonable basis to believe the petitioner is likely to establish
standing to participate in this NRC proceeding, and (2) there is a
legitimate need for access to SUNSI or need to know the SGI requested.
For SGI, the need to know determination is made based on whether the
information requested is necessary (i.e., indispensable) for the
proposed recipient to proffer and litigate a specific contention in
this NRC proceeding \3\ and whether the proposed recipient has the
technical competence (demonstrable knowledge, skill, training,
education, or experience) to evaluate and use the specific SGI
requested in this proceeding.
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\3\ Broad SGI requests under these procedures are thus highly
unlikely to meet the standard for need to know; furthermore, staff
redaction of information from requested documents before their
release may be appropriate to comport with this requirement. These
procedures do not authorize unrestricted disclosure or less scrutiny
of a requester's need to know than ordinarily would be applied in
connection with an already-admitted contention.
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6. If standing and need to know SGI are shown, the NRC staff will
further determine based upon completion of the background check whether
the proposed recipient is trustworthy and reliable. The NRC staff will
conduct (as necessary) an inspection to confirm that the recipient's
information protection systems are sufficient to protect SGI from
inadvertent release or disclosure. Recipients may opt to view SGI at
the NRC's facility rather than establish their own SGI protection
program to meet SGI protection requirements.
7. A request for access to SUNSI or SGI will be granted if:
a. The request has demonstrated that there is a reasonable basis to
believe that a potential party is likely to establish standing to
intervene or to otherwise participate as a party in this proceeding;
b. The proposed recipient of the information has demonstrated a
need for SUNSI or a need to know for SGI, and that the proposed
recipient of SGI is trustworthy and reliable;
c. The proposed recipient of the information has executed a Non-
Disclosure Agreement or Affidavit and agrees to be bound by the terms
of a Protective Order setting forth terms and conditions to prevent the
unauthorized or inadvertent disclosure of SUNSI and/or SGI; and
d. The presiding officer has issued a protective order concerning
the information or documents requested.\4\ Any protective order issued
shall provide that the petitioner must file SUNSI or SGI contentions 25
days after receipt of (or access to) that information. However, if more
than 25 days remain between the petitioner's receipt of (or access to)
the information and the deadline for filing all other contentions (as
established in the notice of hearing or opportunity for hearing), the
petitioner may file its SUNSI or SGI contentions by that later
deadline.
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\4\ If a presiding officer has not yet been designated, the
Chief Administrative Judge will issue such orders, or will appoint a
presiding officer to do so.
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8. If the request for access to SUNSI or SGI is granted, the terms
and conditions for access to sensitive unclassified information will be
set forth in a draft protective order and affidavit of non-disclosure
appended to a joint motion by the NRC staff, any other affected parties
to this proceeding,\5\ and the petitioner(s). If the diligent efforts
by the relevant parties or petitioner(s) fail to result in an agreement
on the terms and conditions for a draft protective order or non-
disclosure affidavit, the relevant parties
[[Page 4256]]
to the proceeding or the petitioner(s) should notify the presiding
officer within ten (10) days, describing the obstacles to the
agreement.
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\5\ Parties/persons other than the requester and the NRC staff
will be notified by the NRC staff of a favorable access
determination (and may participate in the development of such a
motion and protective order) if it concerns SUNSI and if the party/
person's interest independent of the proceeding would be harmed by
the release of the information (e.g., as with proprietary
information).
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9. If the request for access to SUNSI is denied by the NRC staff or
a request for access to SGI is denied by NRC staff either after a
determination on standing and need to know or, later, after a
determination on trustworthiness and reliability, the NRC staff shall
briefly state the reasons for the denial. Before the Office of
Administration makes an adverse determination regarding access, the
proposed recipient must be provided an opportunity to correct or
explain information. The requester may challenge the NRC staff's
adverse determination with respect to access to SUNSI or with respect
to standing or need to know for SGI by filing a challenge within ten
(10) days of receipt of that determination with (a) the presiding
officer designated in this proceeding; (b) if no presiding officer has
been appointed, the Chief Administrative Judge, or if he or she is
unavailable, another administrative judge, or an administrative law
judge with jurisdiction pursuant to 10 CFR 2.318(a); or (c) if another
officer has been designated to rule on information access issues, with
that officer. In the same manner, an SGI requester may challenge an
adverse determination on trustworthiness and reliability by filing a
challenge within fifteen (15) days of receipt of that determination.
In the same manner, a party other than the requester may challenge
an NRC staff determination granting access to SUNSI whose release would
harm that party's interest independent of the proceeding. Such a
challenge must be filed within ten (10) days of the notification by the
NRC staff of its grant of such a request.
If challenges to the NRC staff determinations are filed, these
procedures give way to the normal process for litigating disputes
concerning access to information. The availability of interlocutory
review by the Commission of orders ruling on such NRC staff
determinations (whether granting or denying access) is governed by 10
CFR 2.311.\6\
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\6\ As of October 15, 2007, the NRC's final ``E-Filing Rule''
became effective. See Use of Electronic Submissions in Agency
Hearings (72 FR 49139; Aug. 28, 2007). Requesters should note that
the filing requirements of that rule apply to appeals of NRS staff
determinations (because they must be served on a presiding officer
or the Commission, as applicable), but not to the initial SUNSI/SGI
requests submitted to the NRC staff under these procedures.
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10. The Commission expects that the NRC staff and presiding
officers (and any other reviewing officers) will consider and resolve
requests for access to SUNSI and/or SGI, and motions for protective
orders, in a timely fashion in order to minimize any unnecessary delays
in identifying those petitioners who have standing and who have
propounded contentions meeting the specificity and basis requirements
in 10 CFR Part 2. Attachment 1 to this Order summarizes the general
target schedule for processing and resolving requests under these
procedures.
Dated at Rockville, Maryland, this 13th day of January 2009.
For the Nuclear Regulatory Commission.
Annette L. Vietti-Cook,
Secretary of the Commission.
Attachment 1--General Target Schedule for Processing and Resolving
Requests for Access to Sensitive Unclassified Non-Safeguards Information
(SUNSI) and Safeguards Information (SGI) in This Proceeding
------------------------------------------------------------------------
Day Event/activity
------------------------------------------------------------------------
0..................... Publication of Federal Register notice/other
notice of proposed action and opportunity for
hearing, including order with instructions for
access requests.
10.................... Deadline for submitting requests for access to
SUNSI and/or SGI with information: supporting
the standing of a potential party identified by
name and address; describing the need for the
information in order for the potential party to
participate meaningfully in an adjudicatory
proceeding; demonstrating that access should be
granted (e.g., showing technical competence for
access to SGI); and, for SGI, including
application fee for fingerprint/background
check.
60.................... Deadline for submitting petition for
intervention containing: (i) Demonstration of
standing; (ii) all contentions whose
formulation does not require access to SUNSI
and/or SGI (+25 Answers to petition for
intervention; +7 petitioner/requestor reply).
20.................... NRC staff informs the requester of the staff's
determination whether the request for access
provides a reasonable basis to believe standing
can be established and shows (1) need for SUNSI
or (2) need to know for SGI. (For SUNSI, NRC
staff also informs any party to the proceeding
whose interest independent of the proceeding
would be harmed by the release of the
information.) If NRC staff makes the finding of
need for SUNSI and likelihood of standing, NRC
staff begins document processing (preparation
of redactions or review of redacted documents).
If NRC staff makes the finding of need to know
for SGI and likelihood of standing, NRC staff
begins background check (including
fingerprinting for a criminal history records
check), information processing (preparation of
redactions or review of redacted documents),
and readiness inspections.
25.................... If NRC staff finds no ``need,'' ``need to
know,'' or likelihood of standing, the deadline
for petitioner/requester to file a motion
seeking a ruling to reverse the NRC staff's
denial of access; NRC staff files copy of
access determination with the presiding officer
(or Chief Administrative Judge or other
designated officer, as appropriate). If NRC
staff finds ``need'' for SUNSI, the deadline
for any party to the proceeding whose interest
independent of the proceeding would be harmed
by the release of the information to file a
motion seeking a ruling to reverse the NRC
staff's grant of access.
30.................... Deadline for NRC staff reply to motions to
reverse NRC staff determination(s).
40.................... (Receipt +30) If NRC staff finds standing and
need for SUNSI, deadline for NRC staff to
complete information processing and file motion
for Protective Order and draft Non-Disclosure
Affidavit. Deadline for applicant/licensee to
file Non-Disclosure Agreement for SUNSI.
190................... (Receipt +180) If NRC staff finds standing,
need to know for SGI, and trustworthiness and
reliability, deadline for NRC staff to file
motion for Protective Order and draft Non-
disclosure Affidavit (or to make a
determination that the proposed recipient of
SGI is not trustworthy or reliable). Note:
Before the Office of Administration makes an
adverse determination regarding access, the
proposed recipient must be provided an
opportunity to correct or explain information.
205................... Deadline for petitioner to seek reversal of a
final adverse NRC staff determination either
before the presiding officer or another
designated officer.
A..................... If access granted: Issuance of presiding officer
or other designated officer decision on motion
for protective order for access to sensitive
information (including schedule for providing
access and submission of contentions) or
decision reversing a final adverse
determination by the NRC staff.
A + 3................. Deadline for filing executed Non-Disclosure
Affidavits. Access provided to SUNSI and/or SGI
consistent with decision issuing the protective
order.
[[Page 4257]]
A + 28................ Deadline for submission of contentions whose
development depends upon access to SUNSI and/or
SGI. However, if more than 25 days remain
between the petitioner's receipt of (or access
to) the information and the deadline for filing
all other contentions (as established in the
notice of hearing or opportunity for hearing),
the petitioner may file its SUNSI or SGI
contentions by that later deadline.
A + 53................ (Contention receipt +25) Answers to contentions
whose development depends upon access to SUNSI
and/or SGI.
A + 60................ (Answer receipt +7) Petitioner/Intervenor reply
to answers.
B..................... Decision on contention admission.
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[FR Doc. E9-1152 Filed 1-22-09; 8:45 am]
BILLING CODE 7590-01-P